Recovery of uranium values from residues



tasted REQQVERY @F URANEUM VALUES FRGM RESHDUES No Drawing. ApplicationMay 20, 1952 Serial No. 288,960

13 ill-aims. (Cl. 23-4145) The present invention relates in general torecovery of uranium values from difiicultly-soluble residues, and moreparticularly to an improved method for dissolvably liberating uraniumvalues entrapped in oxide residues which resist repeated leaching withstrong acids.

As is known, in certain uranium processing operations, there obtaindifficultly-soluble residueslargely, insoluble metal oxides, sulfates,and the like-containing, entrapped therein, uranium values which aredesired to be separated and recovered. Commonly, uranium compounds,particularly more or less corrosive compounds such as uranous chloride,upon contacting ordinary processin apparatus, become somewhatcontaminated with materials constituting the apparatus. Most often,these contaminants appear in subsequent stages of processing asresidues1nainly metal oxides-which deposit from aqueous solutions of theuranium. While such residues normally occur in only rather limitedproportion, and are readily removable from solution by filtration or thelike, they often contain small but significant quantities of uraniumvalues. it is frequently highly desirable, particularly in sizableoperations wherein large quantities of such residues continuouslycollect, to recover substantially all of the uranium values therefrom.Although conventional application of strong mineral acid leaching willordinarily recover a part of the uranium content, it has been foundthat, in general, a point is soon reached where further leaching of theresidue with acid removes no more of the uranium therein. From this, itappears that the remaining uranium is mechanically occluded within therefractory residues, thereby rendering further recovery of valuableuranium especially difficult.

This problem is encountered particularly, for example, in processingoperations in which uranium tetrachloride vapors, within an evacuatedchamber, condense and deposit upon various metal apparatus members, aswell as upon porcelain insulators, glass Vessels, and the like. Such.operations are described in more detail in copend-' ing applications:S.N. 532,159, filed April 21, 1944, Patent No. 2,758,000, issued August7, 1956, in the names of James M. Carter and Martin D. Kamen, forIsotope Enrichment Process; S.N. 532,160, filed April 21, 1944, nowabandoned, in the name of Martin D. Kamen, for Process of RecoveringUranium; and SN. 559,624, in the names of James M. Carter and ClarenceE. Larson, filed October 20, 1944, for Process of Recovering Uranium,now Patent No. 2,855,270, issued Ooctober 7, 1958.

In such operations, the condensed uranium is removed from all of thevarious apparatus parts by washing with hot water accompanied by mildscrubbing. The resulting wash solution may be noted to include someinsoluble apparatus-corrosion-product residue, which, together with anyparticles abraded by the scrubbing, are thereupon removed by filtration.Hydrogen peroxide ordinarily is then added to the filtered wash solutionto oxidize the dissolved uranium values to hexavalent oxidation state,in preparation for subsequent uranium processing; here 2,900,226Patented Aug. 18, 1959 ice again, some residue ordinarily precipitatesfrom solution, and is removed from solution by filtration. Theseresidues, together with smaller amounts of like residues similarlyobtained at later points in the processing, are normally combined, giventhree short nitric acid leaches to recover the readily-dissolvableportion of their uranium content, and then dried and accumulated forsubsequent periodic recovery of substantially all of the remaininguranium content. While such residue is, on the whole, a more or lessindiscriminate conglomeration of insoluble inorganic compounds,analytical studies indicate that composition tends largely along thelines of the typical analysis tabulated in Table 1 below.

Probable Form Element Percent by Weight SiOz. @1804. S1102. F9203.OrzOs. A1203. OuO. AgCl. (oxides).

Similar difiieultly-soluble uranium-bearing residues are obtained byvarious other uranium processing operations, as, for example, in thevapor-phase chlorination of uranium oxides to UCL; with carbontetrachloride in ferrousmetal autoclaves, such as is described incopending application: S.N. 737,156, filed March 25, 1947, in the namesof John L. Patterson and Alan Bell, for Uranium Chlorination Process,now Patent No. 2,756,124, issued July 24, 1956.

Upon washing the CCl -free U01 from the autoclave, a uranium-bearingresidue of contaminating metal oxides frequently deposits from the washsolution.

In attempting the desired substantiallycomplete re covery of uraniumtherefrom, such accumulated residue has proven to be strongly resistantto further uranium dissolution. Even most radical dissolutiontreatment-a series of protracted leachings with hot concentrated mineralacid-wil' altord removal of only a part of the remaining uranium; beyondthis, leaching, no matter how vigorous, has proven distinctlyinelfective for the essentially quantitative recovery of the uraniumcontent. It seems clear that the remainder of the uranium is so bound inthe refractory residue-through mechanical co-precipitation, throughadsorption, or perhaps through precipitation of residue particles aroundundissolved uranium compounds-that the uranium is retained effectivelyinaccessible to the acid. In View of the inability of conventionalleaching operations to afford complete recovery of such uranium values,it has become highly desirable that improved methods be found for simplyand quickly overcoming the present, indefinite-occlusion difiiculty, forsubstantially solubilizing the uranium content of the residues, and foraffording substantially quantitative dissolution-recovery of the uraniumfrom such residues.

Accordingly, one object of the present invention is to provide a new andimproved method for the liberation, soluhilization, and recovery, ofuranium values from difficultly-soluble inorganic residues.

Another object is to provide such a method which atfords recovery ofsubstantially the entirety of the uranium content of such residues.

Still another object is to provide such a method which is simple anddirect, and which is especially eiiective in with: HF, then H thenfinally HF.

ANA

rendering uranium values, unleachably occluded in difficultly-solubleinorganic residues, substantially accessible to leaching acids.

A further object. is to provide such .a method constituting 'a generalprocedure of universal applicability for treatment of diverse,indiscriminate, difiicultly-soluble inorganic residues for the recoveryof uranium values therefrom. v l V Additional objects will becomeapparent hereinafter.

In accordance with the present invention, an improved method for therecovery of uranium values from difficultly-soluble inorganic residuecontaining the same, particularly residue comprised predominantly ofmetal oxides, comprises subjecting said residue to the actions, inconsecutive alternation, of gaseous hydrogenfluoride, and'of gaseoushydrogen, thereby diss'olvably liberating uranium values therein, andthereafter leaching the resulting residue with a strongmineral acid. Ithas been empirically discovered that such a series of consecutivealternate treatments with gaseous hydrogen fluoride, and with gaseoushydrogen, is exceptionally eifective in readily rendering uranium valuescontained in said residues virtually completely recoverable by thesubsequent strong acid leaching. For example, when applied to suchresidues as those alluded to in Table 1, supra, it has been found thatthe application of only two gaseous hydrofluorinations, with a singleintermediate hydrogen treatment, followed by aqueous nitric acidleaching, consistently affords recovery of all but so little as 1%, andoften much less, of the initial uranium content of the residue, ascompared with the most vigorous concentrated nitric acid leachingsalone, which regularly leave at least 20% to 50% 'of the uraniumunrecovered. By virtue of such quantitative efliciency, the presentprocess clearly affords significant practical advantages in the recoveryof uranium from residues.

For full effectiveness, the vapor phase treatment should be conducted atelevated temperatures and under substantially anhydrous conditions. 500to 700 C. is the preferred temperature range; hydrogen fluoridetreatment is best effected at GOO-700 C., while 500-600 C. appearsoptimum for'the hydrogen treatment. The vapor phase treatments may mostsimply be effected by passing the gases over the heated residuecontained in boats disposed in a horizontal tube in a tube furnace.Other conventional gas-solid reaction techniques, for example, tumblingthe finely divided'residue in a rotating, heated tube, through which thetreatment gases are passed, are

also effective.

The vapor phase operation should, of course, comprise at least onehydrogen fluoride, and one hydrogen, treatment. It has been found,though, that uranium recovery is somewhat more complete when thealternation ends with a hydrogen fluoride treatment. Similarly, hydrogentreatment appears to be more effective when it follows a hydrogenfluoride treatment. Applying these findings, the simplest alternationsequence for maximum uranium recovery comprises the succession oftreatments recoveries attainable are only slightly lower uponappiication of a sequence comprising only a single treatment with eachgas, in either order.

Concerning the duration of vapor treatments, ordinarily the hydrogenfluoride initially produces a vigorous, exothermic reaction, whichproceeds for only a short time, usually not more than minutes; suchrapid reaction is normally attended with considerable loss of weight ofthe residue, and volatile fluorides, such as silicon tetrafluoride, arenoted to be evolved. Thereafter, the HF reaction becomes much slower,with much smaller progressive weight loss and discontinued fluorideevolution. The reaction with hydrogen is also exothermic, but tends tobe rather slow in reaction rate. To achieve maximum effectiveness inpromoting uranium recovery, the alternated vapor-phase treatment shouldbe Nevertheless, the

continued until completion of reaction is indicated by cessation offurther change of weight of the residue. This is most simply eifected'bycontinuing each treatment in the series until no further change inweight is noted. Applying, in this manner, the aforementioned three-stepalternation to residue of composition as in Table 1, has been found torequire three hours for full completion of the first HF treatment, fourhours for completion of the intermediate H treatment, and again threehours for the final HF treatment. However, further investigation hasrevealed that treatment of the residue by either of the two gaseousreactants is markedly accelerated when it follows a prior treatment ofthe residue with the other gas. For example, while hydrogen reacts withfresh residue very slowly, it proceeds quite rapidly when it follows aprior HF treatment. HF likely renders the residue more porous and/ orconverts various components therein to a form more susceptible toreaction with hydrogen. Likewise, HF engages in a considerably fasterreaction when it follows H treatment. As a result, total alternategaseous treatment time is reduced by employing a multiplicity of short,choppy alternations between HF and H rather than continuing each gaseousreaction to completion before commencing the next treatment. Forexample, it has been found that by employing short, 20 minutealternations, the same maximum weight loss is obtained in 1% hours, asotherwise requires 10 hours using 3 to 4 hour alternations. In thisconnection, though, it is ordinarily preferable that the vaportreatments, even though short, he conducted in consective alternation,rather than simultaneously; generally, mixing the treatment gases hasbeen noted to seriously impair overall uranium recovery efiiciency.Nevertheless, alternation of treatment with one gas with treatment withboth simultaneously-for example, HF, followed by mixed HF+H has beenfound effective.

In large-scale production application it has proven most practical, ineffecting the vapor-phase reaction, to compress the residue in smallpellets and to thereupon pass the gases, in rapid flow, upward throughheated vertical beds of the pellets. Most finely-divided residues arereadily pelletized by moistening with about 10 to 20% water, and thencompressing into discs of ca. 1 inch diameter and about /8 inch to A2inch thickness. The pellets may, for instance, be disposed as a bed of 8to 16 inch depth, on a /8 inch mesh nickel grate, disposed in a lengthof 4 inch Monel pipe surrounded by a vertical tube furnace. Upon flowingHP or H vertically upward through the bed, normally no more than'an houris required for the first treatment reaction to penetrate fully to thecenter of the pellets, with much less time being required in subsequentalternated treatments.

Following the alternating vapor treatment, the residue is leached torecover the uranium values therein. Soluaccordance with the presentinvention are subject to wide 'variation. Among the mineral acids,nitric acid in aqueous solution, especially in 1:1 acid-water weightratio, is particularly preferred. It has further been found that theincorporation of aluminum ions in the leaching acid tends tosubstantially enhance liberation and dissolution of uranium values. Thealuminum ions serve to complex fluoride ions, thereby promotingdissolution of those various residue components, as well as desireduranium values, which have been converted to fluorides by the vaporphaseHF treatment. When employed in conjunction with aqueous HNO as theleaching acid, the preferred source of aluminum ion is dissolvedaluminum nitrate; optimum compositions approximate 1:1 HNO containingabout 15% (by weight) Al(NO Heating, and preferably boiling, the acidthroughout leaching is further beneficial in promoting maximum uraniumrecovery. A series of leaches normally 3 are enough-of about 1 houreach, appears to be the optimum leaching-procedure; ordinarilyaluminumions need be employed in only the first one or two leaches, withsimple aqueous nitric acid being satisfactory for the remainder.Eminently efiiective, as the leaching sequence, is a first leach with1:1 HNO ca. 15% in Al(NO a second leach with 1:1 HNO ca.

. 6 U. As a first set of runs, equal samples of the dried residue. weregiven successive simple leaches with various leaching solutions,including concentrated HNO dilute HNO concentrated HCl, Al(NO solutions,and boil- 10% in A1(NO and a third leach with 1:1 HNO 5 ing water. As asecond set of runs, another group of Recovered uranium values may beremoved and conequal samples were separately treated, in accordance withcentrated from the combined leaching acids, if desired, the presentinvention, to alternate vapor phase treatment by various conventionaltechniques. Solvent extraction with HF, and with H in various sequence,all at 600 C., of uranium with a suitable organic solvent for uraniumfollowed by two Al(NO -1 HNO leaches, and one ions, such as diethyleneglycol dibutyl ether, is satisfac- 10 1:1 HNO leach. In more detail, thevapor phase treattory. In general, before solvent extraction, apretreatments were efiected by disposing each sample in a carbon meritto moderate the rather high acidity of the leaching boat, which wasplaced in a horizontal nickel tube, 2 inch acid should first beapplied-either neutralization with diameter x 29 inch length x inchthickness, running lime, or evaporating off much of the excess HNO andhorizontally through a multiple unit electric furnace. An then dilutingthe remaining solution, is satisfactory. Three atmosphere of nitrogenwas maintained in the tube while extractions with one-half volumes ofthe ether are nor V the sample was brought ,up to temperature, atterwhich w mally'sufficient'for approaching quantitative uranium rereactantgases, in alternating sequence as specified, were moval rather closely.Thereafter, uranium values may be continuously passed through the tube.Thereafter, the quantitatively stripped from the ether with water, saytreated residue was removed from the boat to a glass by employing threeor four one-half volume Water Washes. V l, and leached in the followingmanner. Further concentration of the stripped uranium may then Firstleach. be effected by adding NH OH to precipitate ammonium 120 gm.A1(NO3)39H2O Ware added, then 120 mL dluranate then be 3 5 to U308 H 0,and thereafter 120 ml. HNO (slowly at Further illustration of thequantitative aspects and prefirst); the system was boiled one hour,diluted with ferred cond tions and precedures of the present method 200mL ZNHNOa, boiled 14 hour longer and then is provided in the followingspecific examples. Example entrifuge 1 demonstrates generally theimprovement aflorded by the Second leach: present method for recoveringuranium values from dlffi The remaining residue was boiled with 50 gm.cultly-soluble residues, as compared with conventional A1(NO3)3 9H2Odissolved in 200 m1 1:1 HNO3 leachmg Wlth mmeral for 1 hour, and thencentrifuged.

EXAMPLE 1 Third leach: A large batch of accumulated residue, ofderivation and g g i wf 3 5? g g hour wlth composition substantially thesame as that defined in an en tere Table 1, supra, was obtained. Thismaterial, upon its Analyses were made to determine percentage ofuraremoval from uranium plant processing solutions by filnium recoveredin each case; the various treatments are tration, had been given threenitric acid leachings, dried outlined, and the results of analyses aretabulated, in and stored. Chemical analysis of the batch gave 0.28%Table 2 below.

TABLE 2 Recovery of uranium values from residues SIMPLE LEACHING SampleUranium Product Uranium Run Weight Analysis Treatment Weight Not N0.(gm) (percent) (gm) Removed (percent) lst. oonc. HNO3 leach (1 hr.) 29.5 25.2 A 1 0 28 2nd. conc. HNO; leach (1lir.) 25.6 20.9 3rd. cone. HNOsleach (1 hr) 24. 8 17. 4 4th. conc. HNO; leach (1 hr 24. 0 18. 0 1st.1:1 HNO: leach (1 hr.) 34. 0 27.1 A-2 50 0.28 {2110. 1:1 HNO; leach (1hr.) 25. 7 20.0 3rd. 1:1 BNO; leach (1 hr.) 25.1 19.5 1st. cone. H01leach (1 hr.), 37.8 21. 5 A.i. 50 0. {2nd cone. HCl leach (1 hr.) 32. 316.0 3rd. cone. HCl leach (1 hr.) 29.0 14. 2 A4. 50 0.28 Al(Na)a.9Hzo+l0% H20 (1 11 33.0 30.0 A5 50 0.28 Boiled 1 hr. in H2O 40. 7 v0 PRESENT PROCESS Vapor phase:

(a) HF, 4 hrs. at 600 C 43. 7 ([2) H2, 4lirs. at 000 0 Leaching treatedresidue: B-i 100 0.28 1st. AI(NOa)3-l:l HNO; leach boiled; 1% hrs. 6 5 04 2nd. Al(NO )3-l:l HNO3 leach boiled; 1 hr. 3rd. 1:1 HNO, leachedboiled; 1 hr Vapor phase: o fitfr fiis if go c::::::::'"::: Leaching:(Same as in B-1) 2. 5 0 6 Vapor phase:

HF, 3 hrs. at 600 C 42 8 3-3.- 100 0. 28 H2, 4 hrs. at 000 C.

HF, 3 hrs. at 600 0... Leaching: (Same as in B-l) 3. 0 0. 2

' .Theeflicacy of the present process in afiording virtuallyquantitative uranium recovery, and the improvement afiorded overconventional leaching operations alone, is clearly. evidencedflby theresults set forth in Table 2. Thepreferability of employing theparticular vapor-treatment sequence, HF'H IHF, is also apparent. Theeffectiveness of the particular leach treatment employed after the-vaportreatment is further investigated in the following Example 2.

EXAMPLE 2 The procedure of run B-3 of Example 1 Was repeated separatelyuponeach of several equal samples of residue from the same batchsourceas in Example 1, withthe exception that different leach treatmentswere applied in each case. The vapor-treated residues were givendifferent amounts of leaching, employing'one or more of the individualleaching steps of run 3-3, as indicated. Results are presented in Table3 below.

TABLE 3 Efiect of variation in leaching ing temperature of the acidsolution, reached 135 9 C., and were thereupon dilutedwith an equalvolume of Water. A small amount of precipitate hadformed and wasfiltered on and washed; radiometric determination indicated theprecipitate contained 0.66% of the total uranium present. {The volume ofthe filtrate was 975 ml., and had a density of 1.45, and pH of 0.3. Thesolution was equilibrated three times, in succession, each time withone-half its volume of fresh diethylene glycol dibutyl ether, by meansof shaking in a separatory funnel, permitting phase stratification, anddraining the aqueous phase away from the organic. The separated etherphases were combined, and were washed four times, in succession, eachtime with a volume of water one-half that of the ether phase, applyingconventional separatory-funnel procedure. A fifth water-wash waseffected and tested qualitatively, but showed no trace of furtheruranium in the ether phase. Uranium was then precipitated from thecombined first four water washings by adding ammonium hydroxide SampleUranium Run weight analysis No. (gm.) (percent) Treatment ProductUranium weight not removed (percent) Vapor phase:

HI, 3 hrs. at 600 0. H2, 4 hrs. at 600 C. BF, 3 hrs. at 600 0.

Leaching: O1 100 0.28 (a) Al(NOs)3-1Il HNOs, boiled 1 hr., diluted,boiled }4 hr.

(a) Al(NO )a-1Il HNO; boiled (3-2,.-. 100 0.28 1 hr., diluted, boiledhr.

(1;) 1:1 HNO3, boiled 1 hr. (a) Al(NO3) 1:1 HNOa boiled- 1 hr., diluted,boiled hr. (b) Second A1(N0a)31:1 HNOs leach (as above) (c) 1:1 HNOs,boiled 1 hr.

The results in Table 3 illustrate the preferability of leaching thevapor-treated residue with apair of one hour Al(NO l: 1 HNO boilingleaches, followed by a simple, boiling, aqueous nitric acid leach. InExample 3, following, application of the present method to a largerquantity, and different-type, of residue, as well as organic solventextraction of the uranium from the leaching acid,

followed by stripping with water, ammonium diuranate precipitation, andignition to U 0 are demonstrated.

EXAMPLE 3 Run X .--5 00 grams of dried residue, obtained from the samebatch source as in Example 1, was treated by the procedure outlined forExample 1, run B-3, but employing proportionately greater quantities ofleaching acids. Thereafter, combined filtrates from the leachingprocedure, totaling 1500 ml., were evaporated until the boiltheretountil no further precipitate appeared; the precipitate was removed byfiltration and ignited to U 6 Run Y.A batch of uranium-bearing residue,accumulated at a different point in a uranium processing plant, andhaving a composition somewhat different from that defined in Table 1,was obtained. This material originally had an initial dry weight ofabout 68,000 grams, and had been given four one-hour leaches withboiling b concentrated nitric acid. This treatment had resulted in aresidue of about 4,500 grams dried weight, containing 0.6% uranium bychemical analysis. A sample of this material was subjected to much thesame treatment as outlined in run X above.

Results of these runs, as well as spectrographic analyses of bothresidues, before and after treatment, are set forth in Tables 4 and 5below.

W TABLE 4 Results with varying sample size and composition Run SampleUranium Product Uranium Uranium N0. weight analysis Treatment weightremoved extracted (gm.) (percent) (gn1.) (percent) (gm.)

Vapor phase: 7

HF, 7 hrs. at 650 O 310 Hz, 4}? hrs. at 750 O X 500 0 28 l i g/z hrs. at750 0.- 230 (a) Al(NO3) 1:1 HNOa; 2 hrs (b) Al(NOa)a-1:1 HNOa; 1% hrs 52(0) 1:1 HNO3, 1 hr Vapor phase:

HF 3 hrs. at 600 C. 74 5 Y 100 0 6 Hz, 4 hrs. at 600 O BF, 3 hrs. at 600O Leaching: (Same as in Run X) 34. 5 98. 9 0. 596

TABLE Spectrographic analyses of residues Elements crcent Run No.Material (p Si Ga 511 Fe Or Al Cu Ag Ni Ti Zn Mn Mg Pb Mo Nb X 15.0 10.05.0 2.5 2.5 2.0 1.5 1.0 1.0 1.0 0.6 0.3 0.3 0.3 0.1 0. 45 2. 5 10 7. 510 .04 0. 3 .1 2. 5 .0 0. 1 0. 3 3 0. 1 0. 5 2.0 0.3 10 10 10 .75 .752.5 1.5 1.0 .6 0.3 0.3 .3 1.25 0.6 Y Product:

""""" (a) before leaching 0.1 0. 3 2. 5 10 10 .75 0. 6 2. 5 1.5 .6 0. 30. 3 .3 .1 0. 3 (b) after leaching 01 04 10 2. 5 10 1. 5 0. ()4 08 .1 0.3 .6 1 .04 3 0. 3 2. 0

The exact nature of all of the reactions and mecha-, nisms contributingto the substantially quantitative uranium recovery afforded by thepresent process is not fully known. However, analytical study of thereactions reveals that in the vapor-phase HF treatment, some of theinsoluble residue materials, are, to various extents, converted tofluorides. Much of the silica present is converted to SiF which, beingvolatile, passes away with the treatment gas stream. Similar actionlikely decomposes any silicates present. sulfates, e.g. CaSO also seemto be largely converted to fluorides, with H 80 passing ofl. as vapor.Limited conversion of iron and copper compounds to fluorides appears toobtain also. Upon treatment with hydrogen gas, some of the unconvertedmetal compounds remaining, notably SnO apparently some amounts ofsilver, copper, and iron compounds, and likely some of the convertedfluorides as well, are evidently reduced to free metal. Upon a repeatedtreatment with HF, some of the resulting free metals, and remainingamounts of oxide compounds and the like, are converted to fluorides.Likely, much of the uranium is converted to uranium (IV) fluoridethrough the same reactions. Upon leaching, those compounds converted tofluorides are observed to be most ly acid soluble and dissolve. It isbelieved that all of these eflects mutually contributethrough removal ofcertain residue components by vaporization, removal of others bysolubilization and dissolution, and by changing physical and chemicalconstitution of other components so as to cause expansion orcontraction-to remove part of the residue and render the remainderexceptionally porous, so as to accomplish the virtually completeliberation of the uranium values from the indiscriminate conglomerationof materials in the residue, and render them fully accessible to theleaching acids.

While this invention has been described with particular reference to itsapplication to the processing of appara tus corrosion products,deposited from uranium processing solutions and bearing uranium valuestherein, it is of inherently much wider applicability. The presentprocess is well adapted to improved recovery of uranium unleachablyoccluded in virtually any diflicultly-soluble metal oxide, sulfate orsilicate material, regardless of its source or derivation. A typicalapplication, for example, is in the recovery of uranium values fromores. Another application, of particular current importance, is in therecovery of uranium oxide from spent ceramic fuel elements of hightemperature nuclear reactors; these normally comprise refractorycompositions of thermallystable metal oxides, containing uranium oxidedispersed and bound therein as the nuclear fuel. The uranium content ofsuch fuel elements must be periodically isolated and processed; in suchrecovery, after initial comminution of the oxide elements, the presentprocess has direct application. Various additional applications willbecome apparent to those skilled in the art. It is therefore to beunderstood that all matters contained in the above descripion andexamples are illustrative only and do not limit the scope of the presentinvention.

What is claimed is:

1. In a process for the recovery of uranium values, soluble in a strongmineral acid and contained and bound in compound form within acomminuted solid residue comprised predominantly of at least onematerial selected from the group consisting of inorganic oxides,sulfates, and silicates, comprising leaching such residue with saidstrong mineral acid in which leaching at most only part of the uraniumvalues are recovered, the improvement step for substantially increasingthe degree of completeness of dissolution and recovery of the uraniumvalues from the residue by the leaching, which comprises subjecting saidresidue, prior to said leaching, to the actions, in consecutivealternation, of gaseous hydrogen fluoride, and of gaseous hydrogen,thereby liberating uranium values therein.

2. The process of claim 1 wherein said subjecting of residue to theactions of gaseous reagents is eflected at a temperature maintainedwithin the approximate range of 500 to 700 C.

3. In a process for the recovery of uranium values, soluble in a strongmineral acid and contained and bound in compound form within acomminuted solid residue comprised predominantly of at least onematerial selected from the group consisting of inorganic oxides,sulfates, and silicates, comprising leaching such residue with saidstrong mineral acid in which leaching at most only part of the uraniumvalues are recovered, the improvement step for substantially increasingthe degree of completeness of dissolution and recovery of the uraniumvalues from the residue by the leaching, which comprises, prior to saidleaching, subjecting said residue to the action of gaseous hydrogenfluoride, and then of gaseous hydrogen, and again of gaseous hydrogenfluoride, while maintaining the temperature of the system within theapproximate range of 500 to 700 C., thereby liberating uranium valuestherein.

4. In a process for the recovery of uranium values, soluble in a strongmineral acid and contained and bound in compound form within acomminuted solid residue comprised predominantly of at least onematerial selected from the group consisting of inorganic oxides,sulfates, and silicates, comprising leaching such residue with saidstrong mineral acid in which leaching at most only part of the uraniumvalues are recovered, the improvement step for substantially increasingthe degree of completeness of dissolution and recovery of the uraniumvalues from the residue by the leaching, which comprises subjecting saidresidue, prior to said leaching, to the actions of gaseous hydrogenfluoride for approximately three hours, and then of gaseous hydrogen forapproximately four hours, and then again of gaseous hydrogen fluoridefor approximately three hours, while maintaining the temperature of thesystem substantially within the range of 500 to 700 C., therebyliberating uranium values therein.

5. In a process for the recovery of uranium values, soluble in a strongmineral acid and contained and bound in compound form within acomminuted solid residue comprised predominantly of at least onematerial" selected from the group consisting of inorganic oxides,

only part of the uranium values are recovered, the improvement step forsubstantially increasing the degree of completeness of dissolution andrecovery of uranium values from the residue by the leaching, whichcomprises, prior to said leaching, pelle'tizing said comminuted residueby admixture with about -20% Water and compression, then disposing thepellets so formed as a bed, thereupon subjecting said residue to theaction, in consecutive alternation, of gaseous hydrogen fluoride, and ofgaseous hydrogen, flowed upwardly through said bed, thereby liberatinguranium values therein.

6. In a process for the recovery of uranium values, soluble in a strongmineral acid and contained and bound in compound form within acomminuted solid residue comprised predominantly of at least onematerial selected from the group consisting of inorganic oxides,sulfates, and silicates, comprising leaching such residue with saidstrong mineral acid in which leaching at most only part of the uraniumvalues are recovered, the improvement operation for substantiallyincreasing the degree of completeness of dissolution and recovery of theuranium values from the residue by the leaching, which comprisessubjecting said residue, prior to said leaching, to the actions, inconsecutive alternation, of gaseous hydrogen fluoride, and of gaseoushydrogen, and thereupon effecting said leaching employing aqueous nitricacid as said strong mineral acid.

7. In a process for the recovery of uranium values from a comminutedsolid residue comprised predominantly of at least one material selectedfrom the group consisting of inorganic oxides, sulfates, and silicates,and containing uranium halide bound therewithin, comprising leachingsuch residue with a strong mineral acid in which leading at most onlypart of the uranium values are recovered, the improvement operation forsubstantially increasing the degree of completeness of dissolution andrecovery of the uranium values from the residue by the leaching, whichcomprises subjecting said residue, prior to said leaching, to theactions, in consecutive alternation, of gaseous hydrogen fluoride, andof gaseous hydrogen, and thereupon effecting said leaching employingaqueous nitric acid as said strong mineral acid.

8.111 a process for the recovery of uranium values from a comminutedsolid residue comprised predominantly of at least one material selectedfrom the group consisting of inorganic oxides, sulfates, and silicates,containing uranous chloride bound therewithin, comprising leaching suchresidue with a strong mineral acid in which leaching at most only partof the uranium values are recovered, the improvement operation forsubstantially increasing the degree of completeness of dissolution andrecovery of the uranium values from the residue, prior .to saidleaching, to the actions, in consecutive alternation, of gaseoushydrogen fluoride, and of gaseous hydrogen, and thereupon effecting saidleaching employing aqueous nitric acid as said strong mineral acid.

9. In a process for the recovery of uranium values, soluble in a strongmineral acid and contained and bound in compound form within acomminuted solid residue comprised predominantly of at least onematerial selected from the group consisting of inorganic oxides,sulfates,

and silicates, comprising leaching such residue with said strong mineralacid in which leaching at most only partof the uranium values arerecovered, the improve- 'ment operation for substantially increasing thedegree of completeness of dissolution and recovery of uranium ivaluesfrom the residue by the leaching, which comprises,

prior to said leaching, subjecting such residue to the actions, in.consecutivealternation, of gaseous hydrogen fluoride, and of gaseoushydrogen, and thereuponeffecting said leaching employing strong mineralacid having 11. In a process for the recovery of uranium values,

soluble in a strong mineral acid and contained and bound in compoundform within a'comminuted solid residue comprised predominantly of at,least one. material selected from the group consisting of inorganicoxides, sulfates, and silicates, comprising leaching 's'uch residue withsaid strong mineral acid in' which leaching at most only part of theuranium valuesarerecovered, the improvement step for substantiallyincreasing the degree of completeness of dissolution and recovery of theuranium values from the residue by the leaching, which comprisessubjecting said residue, prior to said leaching, to the actions, inconsecutive alternation, of gaseous hydrogen fluoride, and of gaseoushydrogen, thereby liberating uranium ivalues therein, and thereuponeffecting said leaching by boiling the so treated residue for about onehour with aqueous nitricacid having approximately 15% aluminum nitratedissolved therein, separating the remainingresidue and further boilingit for about one hour with aqueous -nitric acid having approximately5-10% aluminum nitrate dissolved therein, and thereafter againseparating the remaining residue and boiling it for approximately onehour in aqueous nitric acid,.thereby dissolving uranium valuestherefrom.

12. In a process for the recovery of uranium values, soluble in a strongmineral acid and contained and bound in compound within a comminutedsolid residue comprised predominantly of at least one material selectedfrom the group consisting of inorganic oxides, sulfates, and silicatesof an element selected from the group consisting of calcium, iron,copper, tin, and silver, comprising leaching said residue with saidstrong mineral acid in which leaching at most only part ofthe uraniumvalues are recovered, the improvement operation for substantiallyincreasing the degree of completeness of dissolution and recovery of theuranium values from the residue by the leaching, which comprises, priorto said leaching, sub- {jecting said residue to the actions of gaseoushydrogen fluoride for approximately three hours, and then of gaseoushydrogen for approximately four hours, and then again a gaseous hydrogenfluoride for approximately three hours, while maintaining thetemperature of the system .at substantially 600 C., thereby liberatinguranium values therein, and thereupon eifecting said leaching by boilingthe so treated residue for about one hour with aqueous nitric acidhaving approximately 15% aluminum nitratedissolved therein, separatingthe remaining residue and further boiling it for about one hour withaqueous .nitric acid having approximately 510% aluminum nitratedissolved therein, and thereafter again separating the remaining residueand boiling it for approximately one hour in aqueous nitric acid,thereby dissolving uranium values therefrom.

13. In a process for the recovery of uranium values, soluble in a-strongmineral acid and contained and bound in compound form within acomminuted solid residue comprised predominantly of at least onematerial selected from the group consisting of inorganic oxides,sulfates, and silicates, comprising leaching such residue -with saidstrong mineral acid in which leaching at most ium values from theresidue by the leaching, which comprises subjecting said residue, priorto said leaching, to the actions, in consecutive alternation, of gaseoushydrogen fluoride; and of gaseous hydrogen, thereby liberating uraniumvalues therein, thereupon eflecting saidleaching .75 of the so treatedresidue employing aqueous nitric acid 2,900,226 13 M as said strongmineral acid, thereby dissolving uranium References Cited in the file ofthis patent values from the residue, subsequently recovering leacheduranium values from said aqueous nitric acid by solvent UNITED STATESPATENTS extraction with diethylene glycol dibutyl ether, and there-2,227,833 Hixson et a1. J an. 7, 1941 after stripping the extracteduranium values from the 5 2,534,677 Newton et a1 Dec. 19, 1950 otherwith water.

1. IN A PROCESS FOR THE RECOVERY OF URANIUM VALUES, SOLUBLE IN A STRONGMINERAL ACID AND CONTAINED AND BOUND IN COMPOUND FORM WITHIN ACOMMINUTED SOLID RESIDUE COMPRISED PREDOMINANTLY OF AT LEAST ONEMATERIAL SELECTED FROM THE GROUP CONSISTING OF INORGANIC OXIDESSULFATES, AND SILICATES, COMPRISING LEACHING SUCH RESIDUE WITH SAIDSTRONG MINERAL ACID IN WHICH LEACHING AT MOST ONLY PART OF THE URANIUMVALUES ARE RECOVERED, THE IMPROVEMENT STEP FOR SUBSTANTIALLY INCREASINGTHE DEGREE OF COMPLETENESS OF DISSOLUTION AND RECOVERY OF THE URANIUMVALUES FROM THE RESIDUE BY THE LEACHING, WHICH COMPRISES SUBJECTING SAIDRESIDUE, PRIOR TO SAID LEACHING, TO THE ACTIONS, IN CONSECUTIVEALTERNATION, OF GASEOUS HYDROGEN FLUORIDE, AND OF GASEOUS HYDROGEN,THEREBY LIBERATING URANIUM VALUES THEREIN.